The invention relates to the extraction of plutonium and americium-241 from alkaline waste solutions to yield solutions containing a total of less than 10 nanocuries per gram (nCi/g) alpha radiation from plutonium and americium.
A plutonium reclamation facility (PRF) is used to receive and store all sorts of unirradiated plutonium metallurgical scrap such as alloys, metal, compounds, etc., and thereafter to process the scrap to recover and purify plutonium values. In the usual plutonium reclamation facility operation, plutonium bearing scrap or plutonium containing material is dissolved in mixtures of nitric acid and hydrofluoric acid. Aluminum nitrate is thereafter added to complex the residual fluoride and to provide salting strength and facilitate recovery of the plutonium.
The plutonium is recovered by solvent extraction with a 20% tri-n-Butyl-phosphate-carbon tetrachloride solvent. Subsequently, a 30% dibutyl-butyl-phosphonate-carbon tetrachloride solvent extraction process is used to recover most of the residual plutonium and 50 to 60% of the americium-241 from the aqueous raffinate of the tri-n-Butyl phosphate extraction scheme. The aqueous waste from the dibutyl-butyl-phosphonate process, still containing small but significant amounts of plutonium and americium-241, is diluted with large volumes of waste water to form what is termed the plutonium reclamation facility salt waste. This salt waste may be heated to from 110.degree. C. to 120.degree. C. to evaporate the water and dry the salts. The resultant product salt waste contains, ordinarily, from 1000 to 2000 nCi/g of alpha emitters.
The plutonium reclamation facility at Hanford, Wash. may generate in normal operation about 120 cubic meters of salt waste solution per month, which salt waste may typically contain 1.3 molar (M) sodium nitrate, 0.2 M aluminum nitrate, 0.1 M nitric acid, 0.005 to 0.02 M each of ferric ion, magnesium ion and calcium ion, 150 to 300 microcuries per liter (.mu.Ci/l) of americium-241, and 20 .mu.Ci/l of plutonium.
Under current Federal regulations, materials contaminated with transuranic elements in excess of 10 nCi/g cannot be disposed of in normal burial grounds and require storage in a retrievable posture, possibly for periods to 20 years. Large volumes of plutonium containing waste solutions also result from scrap recovery and other plutonium processing operations at other Government installations. These waste solutions may also contain americium-241. Commercially operated plutonium fuel fabrication and/or scrap recovery plants may also produce actinide-containing liquid wastes requiring treatment as well as storage.
Plutonium and americium have been removed from these liquid waste streams by processing these streams through multiple precipitations of iron hydroxide and, in some cases calcium phosphate, at a pH of at least 11, followed in some cases, by a subsequent treatment procedure such as reverse osmosis or passage through beds of bone char. Unless multiple iron hydroxide precipitation are used, the solution resulting from the iron hydroxide precipitation step contains concentrations of plutonium and americium in the scavenged waste solution which are still well above the maximum permissible concentrations (0.004 .mu.Ci/l) of either americium-241 or plutonium-239 in water.
Drawbacks to using the iron hydroxide precipitation process include the necessity to filter and dry large amounts of gelatinous solids after each precipitation step, the generation of relatively large volumes of dried actinide bearing solids requiring storage as actinide wastes, and the need to eventually dispose and immobilize large volumes of dried iron hydroxide. An alternative to using multiple hydroxide precipitation steps would be to store the high salt content, aqueous waste solution from the plutonium reclamation facility having greater than 10 nCi/g of both americium and plutonium into underground tanks. The disadvantages of this disposal method is that a large amount of nonactinide fission product waste become contaminated with actinides thereby increasing the volume of wastes that need to be safely contained. While concentration and drum drying equipment may be used to convert the waste to a solid form which can be subsequently retrievably stored in 50 gallon drums or the like, this would result in a very large volume of actinide (greater than 10 nCi/g) waste requiring expensive, safe interim storage and eventual conversion to another form suitable for terminal storage. In addition, there are engineering difficulties associated with design and remote operation of a drum dryer as well as there is a need for a fine control over the aluminum to sodium ratio to get a satisfactory solid for terminal storage. Because of the above, simpler, more efficient methods for managing plutonium recovery facility waste are required which concentrate actinides to reduce the volume of alpha waste requiring long term storage.
An objective of this invention is to provide a precipitation and ion exchange procedure for reducing the actinide concentration of the waste, when solidified, to preferably below 10 nCi/g. At the present time, the salt waste solution is made alkaline and routed to underground storage tanks where it mixes with other wastes. It is an objective of this invention to avoid converting large volumes of non-actinide waste to retrievable actinide waste (greater than 10 nCi/g).
While sodium titanate has been employed to handle wastes from light-water reactor reprocessing (Chemical and Engineering News, "New Process Consolidates Radioactive Wastes" Vol. 54, No. 2, January 1976, pp. 32-33) this and other references do not address the problem of removing plutonium and americium from high salt content, alkaline waste solutions such as generated in the Plutonium Recovery Facilities.